Nuclear reactors utilize water/steam as a coolant for the reactor as well as a source of energy to power steam turbines to thereby provide electrical energy. Nuclear reactors typically have their nuclear fissionable material contained in sealed cladding tubes, generally of a zirconium alloy, for isolation of the nuclear fuel from the water/steam. Zirconium and its alloys are widely used as nuclear fuel cladding since they advantageously possess low neutron absorption cross-sections, and at temperatures below about 398.degree. C. (the approximate core temperature of an operating nuclear reactor), are non-reactive and importantly possess high corrosion resistance relative to other metal alloys in the presence of de-mineralized water or steam. Two widely used zirconium alloys ("Zircaloys") are "Zircaloy-2" and "Zircaloy-4", trade names of Westinghouse Electric Corporation for zirconium alloys of the above chemical compositions. Zircaloy-2, a Zr-Sn-Ni-Fe-Cr alloy, is generally comprised (by weight) of approximately 1.2-1.7% tin, 0.13-0.20% iron, 0.06-0.15% chromium and 0.05-0.08% nickel. Zircaloy-4 has essentially no nickel, and about 0.2% iron, but is otherwise substantially similar to Zircaloy-2. Zircaloy-2 has enjoyed widespread use and continues to be used at present in nuclear reactors. Zircaloy-4 was developed as an improvement to Zircaloy-2 to reduce problems with hydriding, which causes Zircaloy-2 to become brittle when cooled to ambient temperatures (i.e. when the reactor is shut down) after absorbing hydrogen at higher temperatures.
Zirconium alloys are among the best corrosion resistant materials when exposed to steam at reactor operating temperatures (less than 398.degree. C., typically 290.degree. C.) in the absence of radiation from nuclear fission reactors. The corrosion rate in absence of neutron bombardment is very low and the corrosion product is a uniform, black ZrO.sub.2 oxide film/layer which forms on exterior surfaces of Zircaloy exposed to high temperature steam (uniform corrosion). The black oxide layer of ZrO.sub.2 usually contains a small (non-stoichiometric) excess of zirconium, and as such, it contains excess electrons giving it a black or gray color. It is also highly adherent to zirconium or Zircaloy surfaces exposed to steam.
Despite such relatively high corrosion resistance, when Zircaloys are used as cladding and exposed to high neutron flux in nuclear reactors, corrosion rates are generally increased, and cladding corrosion does become a potential problem in Pressurized Water Reactors (PWR's) and particularly Boiling Water Reactors (BWR's), where corrosion occurs in two formats, namely increased uniform corrosion as mentioned above, and also nodular corrosion. Nodular corrosion is a highly undesirable, white, stoichiometric ZrO.sub.2 oxide layer ("bloom") which forms on the surface of the cladding. It tends to form as small patches ("nodules" or "pustules") on the surface of Zircaloys. Today, it is increasingly common to operate nuclear reactors at high "burn-up" (i.e. to nearly complete consumption of the nuclear fuel). Under these conditions, the cladding is exposed to neutron flux for long periods which generally tends to increase the severity of nodular corrosion. Such increased nodular corrosion not only shortens the service life of the tube cladding (since when concentrated nodular corrosion acts in conjunction with certain contaminants--such as copper ions--localized spalling and ultimately penetration of the cladding can occur), but also produces a detrimental effect on the efficient operation of the reactor. In particular, the white ZrO.sub.2, being less adherent than black ZrO.sub.2, is prone to spalling or flaking away from the tube into the reactor water. On the other hand, if the white nodular corrosion product does not spall away, a decrease in heat transfer efficiency through the Zircaloy tube into the water cooling medium occurs when the less-dense white ZrO.sub.2 oxide layer covers an increasingly large portion of the Zircaloy tube exterior surface. Thus, nodular corrosion can become a significant problem for Zircaloy cladding in situations where Zircaloy tube cladding is left in the nuclear reactor for longer periods in conditions of high "burn-up".
Zircaloys used in cladding for nuclear fuel rods are generally subject during their manufacture to a variety of heat treatments and anneals during the formation of the tubular cladding. It is known that the various heat treatments and quenching procedures used in forming a Zircaloy billet, and the various anneals and cold-working thereafter to form the Zircaloy tube cladding, all have an effect on the particular Zircaloy tubing's ability to resist nodular corrosion, with some Zircaloys able to withstand nodular corrosion better than others despite both being of identical chemical composition. For example, equiaxed Zircaloy-2, heated to 1010.degree. C. and slow-cooled at a rate of 18.degree. C./hr. to 600.degree. C. and thereafter quenched, exhibits a high susceptibility to nodular corrosion under the standard steam test (510.degree. C., 1500 psig, 24 hr.). Paradoxically, the same material, if simply quenched from 1010.degree. C., or if heated to only 950.degree. C. and cooled at the same rate of 18.degree. C./hr. to 600.degree. C. and thereafter quenched, exhibits high resistance to corrosion under the same standard steam test.
Since the actual physical transformations which occur within the Zircaloy composition itself during such processes, in particular during annealing and quenching, were largely not understood, it was prior to this invention difficult to predict what a particular Zircaloy sample's resistance to nodular corrosion will be relative to, for example, another Zircaloy alloy of identical chemical composition but having a different heat treatment and annealing history.
It is prohibitively difficult, time consuming, and expensive to employ existing nuclear reactors and the nuclear reactor environment itself as a means of measuring a particular Zircaloy's tubing sample's resistance to nodular corrosion. At the present time, the standard high-pressure steam test (510.degree. C./1500 psig/24 hr.), or a variation of it, is the only practical means of evaluating the susceptibility to nodular corrosion and predicting in-reactor performance. Using such test, a specimen of Zircaloy tubing is exposed to 510.degree. C. steam under 1500 psig for a period of 24 hours. The presence of nodules of white ZrO.sub.2 oxide on exposed surfaces of such Zircaloy indicate that such Zircaloy material would have a low resistance to nodular corrosion if such specimen were to be used in a nuclear reactor under the conditions as mentioned above. Unfortunately, while the standard high pressure steam test is reliable for determining if Zircaloy tubing has a high susceptibility to nodular corrosion and is accurate in indicating which material will definitely, if exposed to high temperature water/steam in a nuclear reactor at the temperatures indicated above, prove susceptible to nodular corrosion, the converse is not necessarily true. In particular, simply because a Zircaloy in the standard high-pressure steam test passes such test (i.e. fails to show susceptibility to nodular corrosion), it is known that such does not necessarily mean that such Zircaloy will have lengthy immunity to nodular corrosion when exposed to high temperature water/steam in the nuclear reactor environment.
For the above reasons, a test procedure which accurately predicts a Zircaloy sample's susceptibility to nodular corrosion without having to actually expose the samples to a neutron flux environment, is clearly needed.